Corrosion, hydriding, zirconium alloys, Carbon. PowerPoint 2007 Pour appliquer ce modèle à une présentation existante : supprimez toutes les diapos de ce document, puis, onglet [Accueil], "Nouvelle diapositive", "Réutiliser les diapositives", dans le volet Office à droite, cliquez sur [Parcourir], "Rechercher le fichier", une fois le fichier sélectionné, clic droit sur une de ses miniatures, "Insérer toutes les diapositives". NB : ne pas cocher "Conserver la mise en forme source" en bas du volet d'insertion. Corrosion, hydriding, zirconium alloys, Carbon. A. AMBARD, C. DOMAIN, A. ASTORG, P.-E. PLANTET, Y. XU, R. FERNANDES, J. ROQUES*, E. SIMONI* EDF R&D, Département MMC, Les Renardières, 77250 MORET LOING ORVANNES France *Institut de Physique Nucléaire d’Orsay, CNRS-IN2P3, Univ. Paris-Sud, Université Paris-Saclay, F-91406 Orsay Cedex, France

Outline Is it reasonnable to extrapolate what is known from corrosion in primary water to ambient temperature? Internal and external zirconia are different can hydrogen picked-up at ambient temperature damage claddings, guide-tubes and grids? What is the preferred lattice site of carbon in zirconium and in zirconia? Can we measure dissolution rates at amb CAST workshop | Lyon, january 2018

Zirconium alloys In EDF’s plants: Zirconium alloys are used as: Cladding: M5 (Zr1Nb – recrystallized), Zircaloy-4 (Zr1,3Sn – Stress-relieved), Zirlo (Zr1Nb1Sn – stress relieved) Guide tube : Z4 (Stress-relieved and recrystallized) M5, Q12 (Zr1Nb0,5Sn0,1Fe - recrystallized), Zirlo (recrystallized) Grids : same materials as guide tube New materials (not yet used) Optimized Zirlo: Zirlo with tin content decreased, partially recrystallized ATF: Cr coated zirconium alloys CAST workshop | Lyon, january 2018

Generalized corrosion Corrosion is a periodic phenomenon Each period is detected by a « layer » of cracks Corrosion is thermally activated ×2 every 20°C in the operational range No dissolution in the primary water Zirconium is detected in particles Wear scars, spalling etc. Bouineau et al. 2005 Zirconium in the nuclear industry CAST workshop | Lyon, january 2018

Oxide layer Cracks, pores in the outer part (after transition) Oxidised SPP (Laves phases and/or beta precipitates) The exact oxidation states of the various elements might evolve with time Lithium and boron also detected (~100 ppm) The layer is heavily stressed: compressive stress 0,5 – 1 GPa Hu 2013 CAST workshop | Lyon, january 2018

Extrapolation at ambient temperature Anionic growth: zirconium mobility in zirconia is low because of the cost of creating a vacancy in the cationic lattice As the kinetics during a period decreases with time, probably a diffusive mechanism But complicated because not really parabolic The most recent theories incorporate electromigration As such, extrapolation at 20°C is risky We know and expect an enlarge effect of electric field (Zr is a valve metal) Diffusion is dominant at high temperature and probably not at ambient temperature Detailed mechanisms depend on the alloy Demonstrated using M5 and Z4 alloys after ion irradiation: Z4 kinetics increases whereas M5 kinetics decreases Verlet, PhD 2015 CAST workshop | Lyon, january 2018

Internal zirconia Internal zirconia is formed when pellet and cladding contact Microstructure is complex and different from the external zirconia Mixture of tetragonal and monoclinic whereas outer zirconia is mainly monoclinic Doped with fissile products coming from the pellet Internal zirconia differs from outer zirconia Phase fraction are different Compositions are different Ciszak et al Journal of Nuclear Materials 495 (2017) 392 - 404 CAST workshop | Lyon, january 2018

Hydriding Corrosion of Zr generates H, part of which is picked-up by Zr The pick-up fraction lies between 10-20% at high temperature irrespective of the alloy Known exception is Z2 at high burnup which can pick-up > 100% (sometimes) Once H solubility is reached, hydrides precipitate H pick-up mechanism is still under debate Most recent theories link it to the conductivity of the oxide Mobile specy is not really known (OH, H, OH-, H+?) CAST workshop | Lyon, january 2018

High pick-up fraction at ambient temperature Tanabe et al. Mater. Res. Soc. Symp. Proc. 1518, Scientific basis for nuclear waste management XXXVII, Barcelona, Spain, 29 september – 3 october 2013. Kato et al Scientific Basis for Nuclear Waste Management XXXVII, Barcelona, Spain, 29 September – 3 October 2013, Mat. Res. Soc. Symp. Proc. 1518 Couet et al. Corrosion Science 119 (2017) 1-13 CAST workshop | Lyon, january 2018

Hydride reorientation phenomenon When a fuel rod is heated (during transportation for example) hydrides dissolve When it is cooled down under internal pressure (fission gas release), hydride precipitate They tend to precipitate perpendicularly to the max. tension stress Radially in that case Hydride reorientation embrittles the alloys It can not occur at ambient temperature No stress No free hydrogen in Zr alloys at ambient temperature (solubility = 0) Bouffioux et al. Topfuel 2013 CAST workshop | Lyon, january 2018

Hydrogen swelling Hydride have a larger molar volume than Zr (lower density) When they precipitate, they generate an internal stress The stress is released by straining the matrix Experiments show that deformation is roughly 0,1% for 1000 ppm Even at this very high concentration, no cracks appear RWMFRD experiments show that Zr alloys pick-up hydrogen at ambient temperature No limitation a priori But cathodic charging is usually limited to ~1000 ppm Such a damaging is unlikely Blat et al. Zirconium in the Nuclear Industry 2007 Blat et al. Zirconium in the Nuclear Industry 1998 CAST workshop | Lyon, january 2018

<<<<<< C formation energies <<<<<< Ef(C/oct. Site Zr) Ef(C / site 8 ZrO2) Carbon prefer to sit in metallic zirconium rather in zirconia Xu et al. Journal of Nuclear Materials 473 (2016) 61-67 CAST workshop | Lyon, january 2018

diffusion coefficient   D0 (m2.s-1) E (kJ) L (µm) Zirconium 6.02×10-7 85 10-6 Zirconia 1.6×10-7 14 0.8 Xu et al. Journal of Nuclear Materials 473 (2016) 61-67 Gras Cast 3.1 CAST workshop | Lyon, january 2018

C Speciation at zirconia surface Zirconia is highly textured It is stress driven Macroscopic surface can be approximated by one crystallographic plane Garnair PhD 2015 CAST workshop | Lyon, january 2018

Dissolution CAST workshop | Lyon, january 2018

Conclusions High temperature corrosion rate should not be extrapolated to ambient temperature Controlling mechanism is probably changing Internal zirconia should be handled seperately from the outer zirconia layer Hydrogen picked-up at ambient temperature is unlikely to damage zirconium hulls It can be explained considering latest theoritical results gathered at high temperatures Ab initio calculations show that C prefers to sit in zirconium rather than in ZrO2 Tentative experiments show that solubility is low (10-8 mol/L). No real sign of dissolution has been detected. CAST workshop | Lyon, january 2018